ML19305C647
| ML19305C647 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 03/04/1980 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19305C648 | List: |
| References | |
| NUDOCS 8003310208 | |
| Download: ML19305C647 (14) | |
Text
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UNITED STATES 37 NUCLEAR REGULATORY COMMISSION o
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.: l WASHINGTON, D. C. 20555
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VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 57 License No. DPR-32 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated December 15, 1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public;
- and, E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's iegulations and all applicable requirements have been satisfied.
8003310'2.O E
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to the license amendment, and paragraph 3.8 of Facility Operating License No. DPR-32 is herehy amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 57, are hereby incorporated in the license.
The ifcensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION r
A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: March 4, 1980 4
+
o UNITED STATES y
g NUCLEAR REGULATORY COMMISSION L
l wash NGTON. D. C. 20555 f
p VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 56 License No. DPR-37 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated December 15, 1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Ccmmission's rules and regulations set forth in 10 CFR Chapter I; 8.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public;
- and, E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
34Pe5TF9-i
2-2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to the license amendment, and paragraph 3.8 of Facility Operating License No. DPR-37 is amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 56, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Cpecifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION l'!! Y A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to Technical Specifications Date of Issuance: Marcli 4,1980 i
i l
l e
ATTACHMENT TO LICENSE AMENDMENT NOS. 57 AND 56 FACILITY OPERATING LICENSE NOS DPR-32 AND DPR-37 DOCKET NOS. 50-280 AND 50-281 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and certain vertical lines indicating the area of change.
Remove Insert 3.1-7 3.1-7 3.1-8 3.1-8 3.1-10 3.1-10 3.1-11 3.1-11 3.1-12 3.1-12 TS Fig. 3.1-1 TS Fig. 3.1-1 TS Fig. 3.1-2 TS Fig. 3.1-2 TS Fig. 3.1-3 4.2-34 4.2-34 1
i
TS 3.1-7 i
3.
The pressurizer heatup and cooldown rates shall not exceed 0
1000F/hr. and 200 F/hr., respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 3200F.
4.
TS Figure 3.1-1 shall be updated periodically in accordance with the following procedures, before the calculated maximum exposure of the vessel exceeds the exposure for which TS Figure 3.1-1 applies. The curve based on 0.25% Cu weld in TS Figure 3.1-2 shall be used to predict the increase in transition tempera-ture based on fategrated power.
a.
If measurements on the most recently examined irradiation specimen show that its data point is above the 1/4T (thick-ness) line of T.S. Figure 3.1-3 then a new line shall be con-structed through the origin such that it is above all the applicable data points. Once T.S. Figure 1,1-3 is revised, T.S. Figure 3.1-1 must be updated, either by a temperature shift, as by T.S. 3.1.B.4c below, or by revising the applica-ble period (EFPY) to match the new transition temperature from TS Figure 3.1-2.
b.
At or before the end of the integrated power period for which TS Figure 3.1-1 applies, the limit lines on the figure shall be updated for a new integrated power period as follows. The total integrated reactor thermal power from startup to the end of the new period shall be converted to an equivalent integrated neutron exposure. The predicted increase in j
transition temperature at the end of the new period shall then be obtained from TS Figure 3.1-2.
Amendment No. 57, Unit 1 Amendment No. 56, Unit 2
TS 3.1-8 c.
The limit lines in TS Figure 3.1-1 shall be moved parallel to the i:emperature axis (horizontally) in the direction of increasing temperature a distance equivalent to the transition tempert ture increase obtained from TS Figure 3.1-2,less the increment used for the end of the present period.
Basis All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to reactor system temperature and pressure changes.(1) These cyclic loads are introduced by normal unit load transients, reactor trips, and startup and shutdown operation. The number of thermal and loading cycles used for design purposes are shown in Section 4.1 of the FSAR. During unit startup and shutdown, the rates of temperature and pressure are limited. The maximum plant heatup and cooldown rate of 1000F/hr. is consistent with the design number of cycles and satisfies stress limits for cyclic operation.(2)
I The allowable pressure vs. temperature is based on a temperature scale relative to the RT The RT is basically the drop weight NDTI of the material, g.
ET as determined by ASTM E208. However, to assure that this value is conservative, and to guard against the possibility that material with low upper shelf toughness, or with a low rate of increase of toughness with temperature, is not properly evaluated, Charpy test s are also performed. If 35 mils lateral R
l Amendment No.'57, Unit 1 L
Amendment No. 56, Unit 2 i
v
TS 3.1-10 available for the core region weld material but on the basis of actual drop weight data on many similar veld materials, plus the actual Charpy values on this material, the drop weight NDTT is estimated to be 00F.
The RT f r the irst two years of operation included a conservative NDT f
0 estimate of the shift in RT caused by radiation of 100 F.
This added to p
the original RT f 00F assu.ed for the welds, gave a reference RT f
NDT NDT 1000F to be used for the first two years of operation, or until the radiation 0
shif t was estimated to be over 100 F.
In examining the data for the rest of the material in the vessel; as well as the properties for the other ferritic components of the reactor system, it is certain that all other materials initially had RT values significantly l
NDT lower than 1000F.
Since the neutron spectra at the samples and vessel inside radius are identical, the measured (RT) g shif t for a sample can be supplied with confidence to the adjacent section of reactor vessel for some later stage in plant life. The maximum exposure of the vessel 's obtainable from the measured sample data by appropriate application cc 1% calculated azimuthal neutron flux variation.
~
During cooldown and steady state, the thermal stress varies from tensile at l
the inner wall to compressive at the outer wall. The internal pressure super-imposes a tensile stress on this thermal stress pattern, increasing the stress at the inside wall and relieving the stress at the outside wall. Therefore, the limiting stress always appears at the inside wall and the limit line has a l'
f Amendment No. 57, Unit 1 i
Amendment No. 56, Unit 2 l
TS 3.1-11 direct dependence on cooldown rate. For heacup, the thermal stress is reversed i
and the location of the limiting stress is a function of heatup rate.
The 1/4T location is considered conservative since the enhanced metallurgical properties of the surface are not used for the determination of NDTT. The E
1/4T location is used for cooldown and steady state and 3/4T location is used for heatup but the 1/4T location is the most restrictive so it will be the i
controlling curve. In additicn, the limiting NDTT for the reactor vessel 5
after operation is based on the NDTT shift due to irradiation. Since the f.
i fast neutron dose is highest at the inner surface, usage of the 1/4T NDTT 5
criterion is conservative'(FSAR Section 4).
The 500F/hr. heacup and cooldown 2
line on TS Figure 3.1-1 boands all limit lines for heatup and cooldown rates up to 50 F/hr. for indicsted temperatures at or below 4400F, and 1000F/hr.
0 E
[
above 4400F. TS Figure 3.1-1 is based on the Standard Review Plan as modified by measured irradiation sample temperature shif ts and appropriate vessel attenuation factors and azimuthal neutron flux variations.
u TS Figure 3.1-1 defines stress limitations only. For normal operation other inherent plant characteristics, e.g., pump parameter and pressuriser heater 6
4.;
capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure ranges.
The heatup and cooldown rate of 100 F/hr. for the steam generator is consistent 0
with the remainder of the Reactor Coolant Systan, 'as discussed in the first il l}
paragraph of the basis. The stresses are within acceptable limits for the 9
R anticipated usage.
q:_
Temperature requirements for the steam generator correspond with the measured
}..
NDT for the shell. The spray should not be used if the temperature difference i.;
between the pressurizer and spray fluid is greater than 3200F. This limit is imposed to maintain the thermal stresses at the pressurizar spray line nozzle below the design limit.
1 :
Amendment No. 57, Unit 1 Amendment No. 56, Unit 2
TS 3.1-12
References:
(1) FSAR, Section 4.1.5 (2) ASME Boiler & Pressure Vessel Code,Section III, N-415 i
(3) ASME Boiler &
- Pressure Vessel Code,Section III, proposed non-mandatory Appendix G2000 (4) 10 CFR 50, Appendix A, G, & H (5) Regulatory Guide 1.99, Revision 1, April 1977,
" Effects of Residual Elements on Predicted Radiatio'n Damage to Reactor Vessel Materials" (6) USNRC Standard Review Plan, Section 5.3.2, 11/29/75,
" Pressure - Temperature Limits" (7) Welding Research Council (WRC) Bulletin 175, "PVRC Recommendation h
on Toughness Requirements for Territic Materials" (8) WCAP - 7924-A, " Basis for Heatup and Cooldown Limit Curves" (9) Surry Reactor Vessel Radiation Surveillance Program WCAP 7723-Surry 1 F
(July,1972), WCAP 8085-Surry 2 (June,1973)
[
(10) Battelle Columbus Laboratories Research Reports for Surry Pressure Vessel Irradiation Capsula Program.
(a) Surry 1====4 nation and analysis of capsule T (June,1975)
(b) Surry 2 ernmination and analysis of capsule X (Sept., 1975) b (11) ASTM: E185-73, E208, & E23 l
(12) Surry T.S. Change 27 CProposed Change 35) 5 (13) Vepco letter to Mr. Robert W. Reid, NRC Chief Operating Reactors Branch 4, of February 15, 1978, Serial No. 081 I
L Amendment No. 57, Unit 1 l
Amendment No. 56, Unit 2
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UPPER PRES 3URIZATION LIMITS FOR HEATUP AND COOLDOWN SURRY UNITS NO. 1 AND 2 2500 II I
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. o 10, e,
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4 6
8 10 s 2
4 6
8 1020 2
550 F NVT (>1MeV) (N/cm )
l.
Curve 1 - 0.20 % Cu base, 0.25T Cu veld l
Curve 2 - 0.25% Cu base, 0.20% Cu veld Curve 3 - 0.20% Cu base, 0.15% Cu weld L
Curve 4 - 0.15% Cu base. 0.10: Cu veld l
Curve 5 - 0.10: Cu base, 0.05% Cu veld Figure 3.1-2. Radiction induced Increcse In Trcns_i.t.ien. Temperature Amendment No. 57, Unit 1 Amendment No. 56, Unit 2
e I,
i T.S. Figure 3.1-3 I
SURRY 1&2
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c.
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+
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l
/4T = 0.605 (OT) (inner vessel vall).
3
/4T = 0.155 (OT)
~
[
4 (valid to full yessel life)
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1:2 l$
t ii NOTE: Slope of line (0) above was determined from T.S. Figure 3.1-2 with a 0.25% Cu veld for 32 EFPY's and a 2800? shift as determined for T.S. Figure 3.1-1. It was calculated as follows:
f 19N/cm2 0 = 3.85x10 1.203x1018:;/(c _ gyp 7) 2
=
32 EFPY 1
Figure valid up to 11 EFPY Amendment No. 57, Unit 1 Amendment No. 56, Unit 2
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TABLE 4.2-1 SECTION F.
VALVE PRESSURE BOUNDARY (Continued)
Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During No.
Category Areas Methods 5-Year Interval 10-Year Interval Remarks 6.7 (Continued)
The support settings of constant and variable spring-type hangers, snubbers and shock absorbers would be inspected to verify proper distribution of design loads among the associated support components.
SECTION C.
HISCELLANEOUS INSPECTIONS 7.1 Materials Tensile and Capsule 1 shall be Capsules shall Capsule 4 shall be removed Irradiation Charpy V removed and examined be removed and and examined after 20 years.
Surveillance notch (wedge at the first region examined after capsules 5-8 are extra open loading) replacement. Capsule 10 years capsules for complementary and Dosimetry 2 shall be examined or duplicate testing, after 5 years.
7.2 Low IIead SIS Visual (See Remarks)
Not Applicable This pipe shall be visually Piping Located inspected at each refueling in Valve Pit shutdown.
7.3 Low Pressure Visual and 100% of blades Not Applicable gm Turbine Rotor magnetic ff particle or
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